Q1. Answer the following:
- (a)) Calculate the macroscopic scattering cross section for natural uranium. Given ρ = 18.9 g/cm³, percentage weight of 235 U in natural uranium is 0.713; the rest being 238 U. The scattering cross sections for 235 U and 238 U isotopes being 15 b and 13.8 b, respectively. (200 words)
- (b)) Distinguish between prompt and delayed neutrons released in fission and discuss their importance. (200 words)
- (c)) The average number of neutrons produced per fission is 2.5. What would happen in a second, if every neutron released in fission causes another fission? Assume a generation time of 0.1 second. (200 words)
- (d) i)) What is the difference between slowing down density and flux of neutrons? (80 words)
- (d) ii)) If the reference neutron energy is 10 MeV, calculate the lethargy at 200 keV, and 0.025 eV? (120 words)
- (e)) Discuss the concept of breeding and doubling time and describe how abundant fertile actinides could be converted to excellent fissile actinides. (200 words)
- Macroscopic scattering cross section (Σ_s) is the sum of (N_i × σ_s,i) for each isotope in a material.
- Prompt neutrons are emitted instantly from fission (~99.3%) and cause rapid criticality if uncontrolled.
- Delayed neutrons are emitted by fission product decay (~0.7%) and are essential for reactor control by extending generation time.
- Neutron population grows exponentially as N = N₀ × νⁿ, where ν is neutrons per fission and n is generations.
Answer: This response comprehensively addresses key concepts in reactor physics, including the calculation of macroscopic cross sections for natural uranium, distinguishing between prompt and delayed neutrons and their importance in reactor control, analyzing neutron population growth in a hypothetical scenario, defining slowing down density and neutron flux, calculating neutron lethargy, and explaining the principles of breeding and doubling time in nuclear reactors. Each section draws upon fundamental...